The present invention relates to a fuel assembly for a nuclear reactor, and more particularly to a fuel assembly structural member, made of zirconium (Zr) alloys, such as a fuel cladding tube, a spacer, a channel box or the like.
Structural members of a fuel assembly for a nuclear reactor are generally made of zirconium alloy. FIG. 2 is a schematic cross-sectional view showing a fuel assembly for a BWR (boiling water reactor). The fuel assembly is composed of a number of fuel rods 1 each having fuel pellets in a cladding tube, a spacer 2 for retaining the fuel rods at a predetermined interval, a channel box 3 for encasing the fuel rods and the spacer, an upper tie plate 4 and a lower tie plate 5 for holding upper and lower ends of the fuel rods 1, and a handle 6 for transportation of the assembly as a whole. The fuel cladding tubes 11, the spacer 2 and the channel box 3 of these structural members are made of zirconium alloy and are assembled by welding.
FIGS. 3 to 5 show welded portions of the fuel rods 1, the spacer 2 and the channel box 3.
As shown in FIG. 3, end plugs 13 are mounted at welded portions 8 on both ends of each fuel cladding to be 11.
As shown in FIGS. 4A to 4D, spacers 2 are classified into two types, i.e., a lattice type and a circular type. The welded portions 8 of the lattice type spacer 2 are joint portions between spacer bars 21 and a spacer band 22, lattice points 23 at each of which the spacer bars 21 intersect with each other, and an overlapped portion of the spacer band 22, as shown in FIGS. 4a and 4B. On the other hand, the welded portions 8 of the circular spacer are contact points of spacer rings 25, contact portions between the spacer rings 25 and a spacer band 22, and an overlapped portion of the spacer band 22, as shown in FIGS. 4C and 4D.
FIG. 5 shows a shape and welded portions 8 of the channel box 3. The channel box 3 is manufactured by coupling and welding two U-shaped, bent members 31 together, so that two weld lines 8 extends longitudinally.
As described above, any one of the structural members has the welded portions.
The zirconium alloy members are used in the reactor water that is held at a high temperature and a high pressure. In general, the zirconium alloy has a high anti-corrosion and a small neutron absorption cross section. These properties of the zirconium alloy are suitable fuel assembly structural members for a nuclear reactor. The well known alloy are as follows: zircaloy-2 (Sn: 1.2 to 1.7%, Fe: 0.7 to 0.2%, Cr: 0.05 to 0.15%, Ni: 0.03 to 0.08, Zr: remainder); zircaloy-4 (Sn: 1.2 to 1.7%, Fe: 0.18 to 0.24%, Cr: 0.05 to 0.15%, Zr: remainder); Zr-1.0% Nb alloy; Zr-2.5% Nb alloy; Zr-3.5% Sn-0.8% Nb-0.8% Mo alloy (Excel alloy); Zr-1.0% Sn-1.0% Nb-0.2 to 0.5% Fe alloy; Zr-Nb (0.5 to 5%) Sn-(0 to 3.0%)-any one (up to 2.0%) of Fe, Ni, Cr, Ta, Pd, Mo and W; and the like.
It should be noted that, in the description, by weight is represented simply by % except for the case where it is necessary to distinguish these expressions.
When the so-called zircaloy, i.e., Zr-Sn-Fe-Cr-(Ni) alloy is used in a boiling water nuclear reactor, there occurs a local oxidation (hereinafter referred to as nodular corrosion). The generation of the nodular corrosion causes a thickness of normal portions of the structural member to be decreased, and at the same time, causes hydrogen generated in accordance with the corrosion reaction to be absorbed into the members, resulting in embrittlement of the members. Since the corrosion phenomenon is developed in accordance with a lapse of time, it is considered that the corrosion of the members is a factor for determining a service life of the fuel assembly under the operational condition of high degree of burn-up in which these members are used for a long period of time. Also, the hydrogen absorption of this alloy is higher than that of Zr-2.5% Nb alloy. Japanese Patent Unexamined Publication No. 58-22364 shows a heat treatment for quenching members from a temperature of .alpha.+.beta. phase or .beta. phase in order to prevent the nodular corrosion. However, even with this method, it is impossible to prevent the nodular corrosion.
In a Zr-Nb alloy containing Nb, if the amount of Nb is increased, the mechanical strength is increased so that the hydrogen absorptivity is lower than that of the zircaloy. In addition, the local corrosion called "nodular corrosion" does not occur. However, as shown in "proceedings of the International Symposium on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors", Myrtle Beach, S.C., August 22-25, pp. 274-294, since the corrosion property in the weld zone and heat-affected zone thereof is low, there is a problem in using the alloy for the welding structural members. Also, a ductility of the alloy is low so that the alloy is inferior in deformability against impact loads and the like.
Japanese Patent Unexamined Publication No. 61-170552 shows a method for producing a plate member and a tubular member made of high corrosion resistance Zr alloy containing Nb of 0.5 to 2.0%, Sn of 1.5% or less and Fe of 0.25% or less. In order to assemble this alloy as a fuel structural member, it is necessary to weld the high corrosion resistance plate and tubular members, so that the anti-corrosion property of the weld zone will be again degraded.
U.S. Pat. No. 3,121,034 shows a method for improving the weldability under the condition that a cold rolling reduction is 50 to 60% and the final heat treatment is performed at a temperature of 550.degree. to 600.degree. C. for 1 to 240 hours, by using Zr-0.5 to 5% Nb alloy, Zr-0.5 to 5% Nb-0 to 3% Sn alloy or quarternary alloy of Zr-0.5 to 5% Nb-0 to 3 Sn-0 to 2% any one of Fe, Ni, Cr, Ta, Pd, Mo and W. However, it would be difficult to perform an intensive working of several tens of percents for the weld structural member.
Japanese Patent Unexamined Publication 47-28594 shows a method for improving an anti-corrosion property by annealing a Zr-Nb alloy member after welding. However, according to the disclosure of the foregoing literature, even if such a heat treatment is performed, the property of the welded portions is not improved.
On the other hand, with respect to the fuel structural member of multi-layers, Japanese Patent Unexamined Publication Nos. 54-45495, 54-59660, 55-164396, 56-76088, 58-195485, 58-199836 and 58-216988 show a method in which a Zr liner layer is provided on an inner surface of a tube to thereby prevent a stress corrosion cracking due to a mutual effect between uranium dioxide and plutonium oxide pallets and the inner surface of the tube. However, this method has no effect for improving the corrosion resistance property of the outer surface of the tube.
In the foregoing prior art techniques, there is no total consideration for properties needed in the structural members for fuel of high degree of burn-up, such as an corrosion resistance property of weld zone, an corrosion resistance property of non-welded portions, a tensile property, and resistivity against hydrogen embrittlement. These properties have been incompatible with each other.